ML17262A130

Semiannual Radioactive Effluent Release Rept Jan-June 1990.

ML17262A130
Person / Time
Site:	Ginna
Issue date:	06/30/1990
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17261B152	List: ML17261B151 ML17262A130 ML17262A131
References
NUDOCS 9009110086
Download: ML17262A130 (34)

Text

SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT R. E. GINNA NUCLEAR PLANT ROCHESTER GAS AND ELECTR1C DOCKET NO. 50-244 JANUARY - JUNE 1990 9005'11008 ADO~6 OCV

~00>>'~

0S000Z<<

PDC gi J

la cy>>

/

TABLE OF CONTENTS

1.0 INTRODUCTION

2.0 SUPPLEMENTAL INFORMATION 2.1 REGULATORY LIMITS 2.2 MAXIMUM PERMISSIBLE CONCENTRATIONS 2.3 RELEASE RATE LIMITS 2.4 MEASUREMENTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY 2.5 BATCH RELEASES 2.6 ABNORMAL RELEASES 3.0

SUMMARY

OF GASEOUS RADIOACTIVE EFFLUENTS 4.0

SUMMARY

OF LIQUID RADIOACTIVE EFFLUENTS 5 ' SOLID WASTE 6.0 LOWER LIMIT OF DETECTION 7 ' RADIOLOGICAL IMPACT 8 ' METEOROLOGICAL DATA 9.0 LAND USE CENSUS CHANGES 10.0 ANNUAL TABULATION OF PERSONNEL EXPOSURE 11.0 LEAK TEST OF SEALED SOURCES 12.0 CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL 13.0 CHANGES TO THE PROCESS CONTROL PROGRAM 14.0 MAJOR CHANGES TO RADWASTE TREATMENT SYSTEMS

l P

LXST OF TABLES Table 1A Gaseous Effluents Gaseous Summation of all Releases Table 1B Gaseous Effluents Continuous and Batch Releases Table 2A Liquid Effluents Summation of all Releases Table 2B =Liquid Effluents Continuous and'atch Releases Table 3 Solid Waste and irradiated Fuel Shipments Table 4 Release Permits Not Meeting LLD Requirements

I PVc

~ Irking lv

~ ~

~ ~ I 1' INTRODUCTION This Semiannual Radioactive Effluent Release Report is for Rochester Gas and Electric Corporation's R.E. Ginna plant and is submitted in accordance with the require-ments of Technical Specification Section 6.9.1.4. The report covers the period from January 1, 1990 through June 30, 1990.

This report includes a summary of the quantities of radioactive gaseous and liquid effluents and solid waste released from the plant presented in the format outlined in appendix B of Regulatory Guide 1.21, Revision 1, June 1974.

All gaseous and liquid effluents discharged during this reporting period were in compliance with the limits of the R.E. Ginna Technical Specifications.

2.0 SUPPLEMENTAL INFORMATION 2.1 Re ulator Limits The Technical Specification limits applicable to release of radioactive material in liquid and gaseous effluents are:

2~1~1 Fission and Activation Gases The instantaneous dose rate, as calculated in the ODCM, due to noble gases released in gaseous effluents from the site shall be limited to a release rate which would yield

< 500 mrem/yr to the total body and < 3000 mrem/yr to the skin if allowed to continue for a full year.

The air dose, as calculated in the ODCM, due to noble gases released in gaseous effluents from the site shall be limited to the following:

(i) During any calendar quarter to < 10 mrad for gamma radiation and to < 20 mrad for beta radiation.

2 ~ 1~2 Radioiodine Tritium and Particulates The instantaneous dose rate, as calculated in the ODCM, due to radioactive materials released in gaseous effluents from the site as radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than 8 days shall be limited to a release rate which would yield <

1500 mrem/yr to any organ if allowed to continue for a full year.

l I

W+

~s

The dose to an individual, as calculated in the ODCM, from radioiodine, radioactive materials in particulate form and radionuclides other than noble gases with half-lives greater than eight days released with gaseous effluents from the site shall be limited to the following:

(i) During any calendar quarter to < 7.5 mrem to any organ.

(ii) During any calendar year to < 15 mrem to any organ.

2 '.3 Li id Effluents The release of radioactive liquid effluents shall be such that the concentration in the circulating water discharge does not exceed the limits specified in accordance with Appendix B, Table II, Column 2 and notes thereto of 10CFR20. For dissolved or entrained noble gases the total activity due to dissolved or entrained noble gases shall not exceed 2 E-4 uCi/ml.

The dose or dose commitment to an individual as calcu-lated in the ODCM from radioactive materials in liquid effluents released to unrestricted areas shall be limited:

(i) During any calendar quarter to < 1. 5 mrem to the total body and to < 5 mrem to any organ, and (ii) During any calendar year to < 3 mrem to the total body and to < 10 mrem to any organ.

2 ' Maximum Permissible Concentrations MPC 2~2~1 For gaseous effluents, maximum permissible concentrations are not, directly used in release rate calculations since the applicable limits are stated in terms of dose rate at the unrestricted area boundary.

2.2.2 For liquid ef fluents, the maximum permissible concen-tration values specified in 10CFR20, Appendix B, Table II, column 2 are used to calculate release rates 'and permissible concentrations at the unrestricted area boundary. A value of 2E-04 uCi/ml is used as the MPC for dissolved and entrained noble gases in liquid effluents.

~ ~

I' h:

II ~

V I h

Ih I ~ 4

/ C 2 ' Release Rate Limits The release rate limits for fission and activation gases from the R.E. Ginna plant are not based on the average energy of the radionuclide mixture in gaseous effluents; therefore, this value is not applicable.

However, the average energy of the radionuclide mixture was 0.285 Mev.

2.4 Measurements and A roximations of Total Radioactivit Gamma spectroscopy was the primary analysis method used to determine the radionucl ide composition and concen-tration of gaseous and liquid effluents. Composite samples were analyzed'or Sr-89, Sr-90 and Fe-55 by a contract laboratory. Tritium and alpha analysis were done using liquid scintillation and gas flow proportional counting respectively.

The total radioactivity in effluent release was deter-mined from the measured concentration of each radio-nuclide present and the total volume of effluents released.

2 ' Batch Releases 2.5.1 ~zi id 1~ Number of batch release: 2.61 E+02 2~ Total time period for batch releases: 4.47 E+04 min 3 ~ Maximum time period for a batch release: 6.15 E+03 min 4 ~ Average time period for batch releases: 1.71 E+02 min

5. Minimum time period for a batch release: 9.0 E+00 min 6 ~ Average stream flow (LPM) during periods of effluent releases into the flowing stream:

a. Effluent Release Flow 4.46 E+02 LPM

b. Dilution Stream Flow 9.94 E+05 LPM 2.5.2 Gaseous 1~ Number of batch releases: 1.4E+01 2 ~ Total time period for batch releases: 2.33E+04 min 3 ~ Maximum time period for a batch release: 3.81E+03 min 4 ~ Average time period for batch releases: 1.67E+03 min 5 ~ Minimum time period for a batch release: 2.44E+02 min

'L 4

4 I

I II 4'

2 ' Abnormal Releases There were no abnormal releases of liquid or gaseous effluents during the reporting period.

3 '

SUMMARY

OF GASEOUS RADIOACTIVE EFFLUENTS The quantities of radioactive material released in gaseous effluents are summarized in tables 1A and 1B.

All releases were considered to be elevated releases.

4.0

SUMMARY

OF LIQUID RADIOACTIVE EFFLUENTS The quantities of radioactive material released in liquid effluents are summarized in tables 2A and 2B.

5~0 SOLID WASTES The quantities of radioactive material released; in shipments of solid waste transported from the site during the reporting period are summarized in table 3.

Principal nuclides were determined by gamma spectroscopy and non-gamma emitters were calculated from scaling factors determined by an independent laboratory from representative samples of that waste type.

6 ' LOWER LIMIT OF DETECTION NOT MET One or more gamma emitting radionuclide did not meet the required lower limit for detection for 1 gaseous release and 24 liquid releases. These are listed by release number in table 4.

7.0 RADIOLOGICAL IMPACT An assessment of doses to the maximally exposed indivi-dual from gaseous and liquid effluents will be performed and reported in the July - December, 1990 Semi-Annual Report for the year of 1990.

8 ' METEOROLOGICAL DATA Not applicable for this report.

9 ' LAND USE CHANGES Not applicable for this report.

10.0 ANNUAL TABULATION OF PERSONNEL EXPOSURE This data will be in the report issued for July-December, 1990.

l

<<4.

LEAK TEST OF SEALED SOURCES No sealed sources were found to be leaking when smeared by both wet and dry smears.

CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL (ODCM)

There were no changes to the ODCM during the report period.

CHANGES TO THE PROCESS CONTROL PROGRAM (PCP)

During the current reporting period, the PCP was changed to include a vendor supplied demineralization system for processing of liquid radwaste. This system utilizes mixed media filtration, anion, cation and mixed bed resin to process water from the waste hold-up tank.

The installation of the demineralization system for processing of liquid radwaste was necessary because the evaporator and recycle systems were not able to effec-tively process the liquid waste due to reduced capacity of the evaporator package; and the onsite storage capability was near capacity and severely limiting operation flexibility.

Section V on spent bead resin processing was further segregated to include sections on primary processing, ef fluent stream processing and shipment preparation.

This was necessitated by the production of spent resin from the vendor supplied demineralization system.

Liquid radwaste processed through a demineralization system results in the production of spent bead resin media. Bead resin is a normal waste form produced and disposed of at Ginna as described in Section V.A.

Therefore, additional production of this waste form will meet existing criteria.

Leakage from the spent fuel pit was added as a source of liquid waste.

A copy of the PCP with the changed sections marked is included with this report.

J'

MAJOR CHANGES TO RADWASTE TREATMENT SYSTEMS During the current reporting period, a temporary, liquid radwaste system was installed to supplement the waste evaporator. The system is a fluidized transfer demineralization system consisting of 5 to 6 resin vessels, booster pump, mechanical filter, dewatering pump and process control unit. The entire system is interconnected with flexible reinforced non-collapsible butyl rubber hoses designed for temperatures between

-20 F and 180 F and pressure from 0 to 300 psig. The supplied system is designed and operated in accordance with the following standards and operating parameters:

a) Reg Guide 1.143 b) ANSI SS.2 c) ANSI/ASME B31 1F d) ASME B+PV Code Section VIII and IX e) Pressure 0-150 psig f) Temperature 50-135 F (resin limited) g) Plow 15-200 gpm h) Hydro tested'to 225 psig The shut-off head of the booster pump and the monitor tank transfer pump is 100 and 115 psig respectively.

This is well below the design of all the temporary system components.

The temporary system processes waste from the waste hold-up tank using one of the monitoring tanks as a batch tank. Provisions have been made to allow the waste to be recycled through the resin beds until chemistry and activity release parameters are met.

When the resin media is spent, shipping container for'isposal.

it is sluiced to a

g I

~ +

f 40 I ~

$p

Table lA KFTXIJENT AND HASTE DISPOSAL SEMIANNUAL REPORT GASEOUS EFFLUENTS SUMMATION OF ALL RELEASES January 1 June 30, 1990 Unit )Quarter )Quarter (Est. Total(

Error A. Fission S activation gases

1. Total release Ci 1.66E+02 4.28E+01 7.0E+00

2. Avera e release rate for eriod uCi sec 2.13E+01 5.44E+00

3. Percent of technical s ecification limit 4 3.38E-03 8.64E-04 B. Iodines

1. Total iodine-131 Ci 2.57E-03 3.10E-03 2.6E+01

2. Avera e release rate for eriod uCi sec 3.31E-04 3.94E-04

3. Percent of technical s ecification limit 7.26E-01 8.67E-01 C. Particulates

1. Particulates with half-lives > 8 da s Ci 7.24E-01 1.16E+00 3.0E+Ol

2. Avera e release rate for eriod uCi sec 9.31E-02 1.49E-Ol

3. Percent of technical s ecification limit 4.95E-06 7.93E-06

4. Gross al ha radioactivit . Ci D. Tritium

1. Total release Ci 3.27E+01 1.66E+Ol 3.2E+00

2. Avera e release rate for eriod uCi sec 4.21E+00 2.11E+00

3. Percent of technical s ecification limit 4.95E-04 2.48E-04

I II 4I Y

+W

Table 1B EETI13ZNT AND HASTE DISPOSAL SEMIAHKJAL REPORT GASEOUS EFFLUENTS - ELEVATED RELEASE CONTINUOUS MODE BATCH MODE Nuclides Released Unit uarter arter uarter arter

1. Fission gases 1 2 1 2 k ton-85 Ci 1.75E-01 1.76E+00 4.58E+00 k ton-85m Ci 4.59E-02 6.67E-03 k ton-87 Ci 4.41E-02 1.20E-03 k ton-88 Ci 8.68E-02 1.53E-02 xenon-133 Ci 1.44E+02 6.97E+00 3.08E+02 2.65E+01 xenon-135 Ci 9.76E+00 2.95E+00 4.28E-02 xenon-135m Ci 1.64E+00 1.36E-01 xenon-138 Ci 7.02E-02 4.70E-02 Others s ecif Ci ar on-41 Ci 2.70E-01 6.91E-02 xenon-131m Ci 3.15E-01 6.04E-03 3.32E+00 1.52E+00 xenon-133m Ci 2.54E+00 9.36E-03 1.99E+00 2.57E-02 Total for eriod Ci 1-.59E+02 1.02E+01 7.12E+00 3.26E+Ol

2. Iodines iodine-131 Ci 1.83E-03 2.21E-03 1.11E-04 8.58E-04 iodine-133 Ci 6.25E-04 2.70E-05 8.40E-07 3.28E-07 iodine-135 Ci Total for eriod Ci 2.46E-03 2.24E-03 1.12E-04 8.58E-04

3. Particulates strontium-89 Ci strontium-90 Ci cesium-134 Ci cesium-137 Ci 2.24E-06 barium-lanthanum-140 Ci Others s ecif Ci cobalt-60 Ci 5.29E-07 carbon-14 Ci 3.74E-01 3.73E-01 3.50E-01 7.84E-Ol unidentified Ci 7.79E-06 5.83E-06 1.53E-07 Note: Isotopes for which no value is given were not identified in applicable releases.

I '1 I I j<<N I

J

Table 2A EFEXIK5T AND WLSZE DISPOEM LIQUID EFFIIJEH'IS - SUMMATION OF ALL RZXZASES January 1 June 30, 1990 I Unit IQuarter IQumter IEst.TotalI Error A. Fission and activation products Il. Total release (not including tritium, ses al ha Ci 4.63E-02 3.34E-02 9.0E+00 I2. Average diluted concentration I I I dur iod uCi ml 3.31E-10 2.81E-10

3. Percent of a licable limit 5.12E-02 1.55E-02 B. Tritium

1. Total release Ci 1.45E+02 8.24E+01 3.2E+00 I2. Average diluted concentration I I I I dur iod uCi ml 1.04E-06 6.92E-07

3. Percent of a licable limit 3.46E-02 2.31E-02 C. Dissolved and entrained gases

1. Total release Ci 2.34E-01 3.37E-02 8.0E+00 I2. Average diluted concentration I I I I dur iod uCi. ml 1.67E-09 2.83E-10

3. Percent of a licable limit: 8.36E-04 1.42E-04 D. Gross alpha radioactivity

1. Total release Ci E. Volume of waste released rior to dilution liters 2.77E+07 2.65E+07 5.0E+00 F..Volume of dilution water used dur'od liters 1.40E+ll 1.19E+ll 5.0E+00

1 t II

~

'II tt k>

Table 2B EFFLUENT AND HASTE DISPOSAL SEMIANNUAL REPORT LIQUID EFFLUENTS CONTINUOUS MODE BATCH MODE Nuclides Released Unit uarter uarter uarter uarter strontium-89 Ci 1.63E-05 strontium-90 Ci 5.27E-06 1.81E-04 cesium-134 Ci 8.19E-04 1.57E-03 cesium-137 Ci 1.75E-06 1.01E-04 4.68E-03 6.86E-03 iodine-131 Ci 9.71E-05 1.47E-06 1.90E-02 3.83E-03 cobalt-58 Ci 1.29E-03 2 '8E-03 cobalt-60 Ci 1.43E-07 5.89E-03 3.25E-03 iron-59 Ci 2.31E-06 9.95E-05 zinc-65 Ci man anese-54 Ci 1.93E-04 2.94E-04 chromium-51 2.85E-03 zirconium-niobium-95 Ci 4.58E-04 6.93E-04 mol bdenum-99 Ci 6.69E-06 technetium-99m Ci barium-lanthanum-140 Ci 2.41E-06 2.37E-04 cerium-141 Ci Other s ecif Ci silver-110m 3 '2E-03 1.10E-03 cesium-136 Ci 2.27E-04 iodine-133 Ci 2.88E-06 3.11E-06 6.94E-03 3.40E-03 iodine-135 1.03E-06 9.69E-06 1.19E-03 3.41E-03 ruthenium-103 Ci 4.22E-05 ruthenium-106 Ci 1.56E-04 antimon -122 Ci 1.10E-04 antimon -124 Ci 4.08E-04 1.32E-03 antimon -125 4.80E-04 5.42E-05 tellurium-131m Ci 2.49E-04 iron-55 Ci 1.17E-03 1.83E-03 unidentified Ci Total for eriod above Ci 1.03E-04 1.16E-04 4.62E-02 3.34E-02 xenon-133 Ci 2.33E-01 3.28E-02 xenon-135 Ci 1.11E-03 8.51E-04 NOTE: Isotopes for which no value is given were not identified in applicable releases.

I t i1

Table 3 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel) 6-month Est. Total

1. T e of waste Unit Period Error

a. Spent resins, filter sludges, m 2. 63 E+01 2E+00 eva orator bottoms etc. Ci 2.16 E+01 5E+00

b. Dry compressible waste, con- m 4.05 E+01 2E+00 taminated e i etc. Ci 2.37 E+00 5E+00

c. Irradiated components, control m rods etc. Ci

d. Other (describe) m Ci

2. Estimate of major nuclide composition (by type of waste)

a. Fe-55 2. 3 E+01 Co-60 1. 9 E+01 Sb-125 l. 3 E+01 Cs-137 8.7 E+00 Co-58 6.8 E+00 Ni-63 1. 8 E+01

b. Fe-55 2. 6 E+01 Co-60 1.5 E+Ol Ni-63 7.9 E+00 Cs-137 2.3 E+01 Co-58 1.9 E+01 Sb-125 5.3 E+00

3. Solid Waste Disposition Number of Shi ments Mode of Trans ortation Destination 25 Highway Vehicle Barnwell, SC B. IRRADIATED FUEL SHIPMENTS (Disposition)

Number of Shi ments Mode of Trans ortation Destination None

Table 4 RELEASE PERMITS NOT MEETING LLD REQUIREMENTS No. Date Isoto es Cause 9000019-G I 4/9/90 (Co-60, Ce-144 a 9000120 L(3/8/90 IFe-59, Zn-65, Cs-134, Ce-141 b 9000136-L I 3/19/90 (Fe-59, Zn-65, Cs-134, Ce-141 b 9000141-L 3/22/90 I (Fe-59, Zn-65, Cs-134, Cs-137, Ce-141 b 9000144-L( 3/23/90 (Fe-59, Zn-65, Ce-141 b 9000145-L 3/23/90

( (Fe-59, Zn-65, Ce-141 b 9000146 L(3/24/90 (Fe-59, Zn-65 b 9000149-L(3/26/90 (Ce-141 b 9000159-L(4/1/90 (Zn-65, Cs-134, Ce-141 b 9000166-L(4/5/90 (Zn-65 b 9000167 L(4/5/90 (Fe-59, Zn-65, Cs-134, Ce-141 b 9000168 L(4/6/90 (Fe-59, Zn-65, Cs-134 b 9000172-L(4/10/90 (Zn-65, Cs-134 b 9000174-L 4/10/90 I (Zn-65, Cs-134 Ce-141 b 9000175-L 4/13/90 I (Zn-65 b 9000177 L(4/17/90 IFe-59 b 9 00017 8-L 4/18/9 0 I (Fe-59, Zn-65 b 9000184 L(4/23/90 (Zn-65 b 9000186 L(4/25/90 (Fe-59, Zn-65 b 9000196 L(4/30/90 (Fe-59, Zn-65 b 9000210-L 5/10/90 I (Fe-59, Zn-65 b 9000233 L 5/24/90 I (

Cs-134 Ce-141 b 9000252 L 6/9/90 I (Fe-59, Zn-65, Ce-141 b 9000253-L( 6/9/90 (Fe-59, Zn-65 b 9000276-L 6/26/90

( IFe-59, Zn-65 b a ~ Grab sample for release was not, large enough to meet the LLD of lE-12 uCi/ml in the count time selected.

b. Activity from other isotopes caused an increased background resulting in the LLD calculation exceeding 5E-07 uCi/ml for the listed isotopes.

0 I GINNA STATION PLANT OPERATIONS REVIEW COMMITTEE MEETING DATE: 90/01/12 MEETING 5 005 MEMBERS PRESENT: MEMBER ALTERNATE CHAIRMAN J. A. WIDAY SUPT. GINNA SUPPORT SERVICES OPERATIONS MANAGER L. F. SMITH MAINTENANCE MANAGER TECHNICAL MANAGER S. T. ADAMS RESULTS & TEST SUPERVISOR G. E. JOSS REACTOR ENGINEER HP & CHEMISTRY MANAGER D. L. FILKINS QUALITY CONTROL ENGINEER I & C SUPERVISOR NUCLEAR ASSURANCE MANAGER MAINT PLANNING/SCHED MANAGER OTHERS ATTENDING R. PLOOF PORC. SECRETARY J. WRIGHT/L. STAVALONE THE CHAIRMAN CALLED THE MEETING TO ORDER.

3.0 REVIEW OF PROCEDURE CHANGES 3.2.0-90-005-001 PT-3 89-4459 CONTAINMENT SPRAY PUMPS AND NaOH ADDITIVE SYSTEM THE PROC. SPEC. PRESENTED THIS PCN:-

ZT REQUESTED THIS PROCEDURE BE DELETED TO ALLOW INCORPORATION OF NEW SPECIFIC MONTHLY AND QUARTERLY TEST PROCEDURES PT-3M AND PT-3Q. THE COMMITTEE REVIEWED AND RECOMMENDED APPROVAL OF THIS PROCEDURE FOR DELETION. THIS CONSTITUTES A PROCEDURE CHANGE THAT MEETS THE EXEMPTION REQUIREMENTS OF THE 10 CFR

PAGE 3 CHANGES OR VIOLATIONS WERE INVOLVED AND THERE ARE NO UNREVIEWED SAFETY QUESTIONS. THIS ITEM 1S COMPLETE.

9 ' OTHER DISCUSSION 9.1.0-90-005-001 SAFETY ANALYSIS FOR TEMPORARY RADWASTE DEMINERALIZER SYSTEM THE TECHNICAL MANAGER PRESENTED THE ATTACHED SAFETY ANALYSIS FOR THE TEMPORARY RADWASTE DEMINERALIZER SYSTEM. PORC REVIEWED THIS ANALYSIS AND PRELIMINARY SAFETY EVALUATION AND CONCLUDED THE PROPOSED TEMPORARY MODIFICATION DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION. THIS ITEM IS COMPLETE.

ALL OF THE ABOVE ITEMS WERE REVIEWED BY THE COMMITTEE WITH RESPECT TO THE TECHNICAL SPECIFICATIONS AND THE COMMITTEE HAS DETERMINED THAT NO TECHNICAL SPECIFICATION CHANGES OR VIOLATIONS WERE INVOLVED IN THE CHANGES AND THERE ARENO UNREVIEWED SAFETY QUESTIONS.

THE CHAIRMAN ADJOURNED THE MEETING.

W GHT/L. STA ALONE PORC SECRETARY APPROVED BY:

S. M. S CTOR T ~ L. ALE ANDER SUPERINTENDENT QC ENGINEER

1

~ ~